ProgramMaster Logo
Conference Tools for 2024 TMS Annual Meeting & Exhibition
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Presentation Title Evaluating the Strength of TRISO Pebbles via Drop Tests and Nondestructive Techniques
Author(s) Assel Aitkaliyeva, Mitchell Mika
On-Site Speaker (Planned) Assel Aitkaliyeva
Abstract Scope Understanding and quantifying the degradation of fuel pebbles is key to qualifying them for use in next generation advanced reactors. In this contribution, we assess the integrity of the pebbles using a combination of drop tests and nondestructive evaluation techniques. The pebbles are dropped from the predetermined height (consistent with existing high temperature gas reactor designs) onto three different surfaces: graphite, concrete, and metal. The integrity of the pebbles after the drop is then assessed using nondestructive techniques such as ultrasound and X-ray tomography. We developed a new and robust testing approach that allows evaluating a large number of samples quickly, which can be easily adapted to use during reactor operations. The results of the drop testing provide a baseline for the drop strength of TRISO pebbles to inform reactor design standards.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Mechanical Properties, Characterization

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced Fuel for Integrating Nuclear SMRs with Renewables
Analyzing ZrN and LiF-doped ZrN as a Shield Material for Fusion and Space Reactors
Aspects of Graphite Performance in Molten Salt
Beryllium Carbide Tolerance to Radiation Damage for Advance Reactor Moderators
Correlating Atomic Structure with Elastic Properties in Non-textured Pyrocarbon
Elucidating the Effect of Radiation-induced Defect Accumulation on Swelling in UN Using In-situ TEM Irradiation
Evaluating the Strength of TRISO Pebbles via Drop Tests and Nondestructive Techniques
Exploring Hydrogen Absorption with High-density Fuels
From First-principles, Modeling the Effect of Point-defect Phonon Scattering on the Thermal Conductivity of Oxide Fuels
Impact of Grain Boundary and Surface Diffusion on Predicted Fission Gas Bubble Microstructural Evolution Behavior and Release in UO2 Nuclear Fuel
In Situ High-temperature 3D Imaging of the Damage Evolution in a SiC Nuclear Fuel Cladding Material
Mechanical and Thermophysical Properties of ZrC, NbC, and TaC Binary Carbide Surrogate Fuels for Nuclear Thermal Propulsion Systems
Microstructure, Mechanical Properties, and Residual Stresses of Cr-Coated SiC Fuel Cladding for Light Water Reactors
Multiphysics and Multiscale Modeling of the Mechanical Properties of the Porous Pyrocarbon Buffer Layer in TRISO Particle Fuel
Multiphysics and Multiscale Modeling of Micro- and Macro- cracking in High Burnup UO2 Fuels
Next Steps in TRISO Fuel Technology Development
Radially Resolved Thermo-physical Modelling in High Burnup Oxide Nuclear Fuel
Radiation Effects and Corrosion of Silicon Carbide in Nuclear Reactor Environments
Radiation Effects in Ceramics for Immobilization of Actinide-containing Nuclear Waste
Reduced Processing Temperature of Advanced Ceramic Composites
Response of the U3Si2 + 50 wt%UB2 Composite Alloyed with Al, Al2O3, Y and Y2O3 in High-Temperature Oxidizing Atmospheres
Steam Corrosion of Cr- and Zr-containing Uranium Nitride Fuels: Mechanistic Insights from In Situ Neutron Diffraction
Thermal Conductivity Suppression in Uranium-doped Thorium Dioxide Due to Phonon Resonant Scattering
Thermo-mechanical Characterization of a Novel Alumina-Yttria Ceramic Coating for Lead Fast Reactors
Thermodynamics of Complex Carbides for Nuclear Applications in Extreme Environments
Thermophysical Properties of Solid Solution Carbide Fuels for Nuclear Thermal Propulsion
Thermophysical Properties of Uranium Nitride-metal Composite Fuels
Unraveling the Influence of Charge Effect on Defect Recombination in ThO2
Uranium Enrichment Homogeneity Study on HALEU/LEU UO2 Fuel Pellets
Using In Situ Neutron Powder Diffraction to Study the Thermal Expansion of Fission Product Doped UN

Questions about ProgramMaster? Contact programming@programmaster.org